1. Field of the Invention
Example embodiment(s) of the present invention are directed to methods of improving the energy output of a nuclear reactor, while satisfying a minimum subcritical bank withdrawal position (MSBWP) safety condition, to a method for determining fuel rods to be subject to an extended natural uranium blanket layer, so as to employ a variable natural blanket for selectable fuel rods in one or more fuel bundle(s) of the reactor, and to a fuel bundle which has a variable natural uranium blanket.
2. Description of the Related Art
FIG. 1 illustrates a conventional boiling water reactor (BWR); FIG. 2 illustrates a fuel bundle in the core of a reactor vessel; and FIG. 3 represents an x-y map distribution of a 10×10 array of fuel rods and water rods in a fuel bundle located in one quadrant of a control blade.
As show in FIG. 1, a pump 110 supplies water via a conduit (e.g., in the direction of arrow from pump 110 towards the containment vessel 114) to a reactor vessel 112 housed within the containment vessel 114. The core of the reactor vessel includes a number of fuel bundles B at locations. Controlled nuclear fission takes place at the fuel bundles (e.g., fuel bundles B of FIG. 2) in the core and generates heat which turns the supplied water (not shown)—flowing between fuel rods—into steam.
Referring to FIG. 1, steam (not shown) is supplied from the reactor vessel 112 to turbines 118 for powering a power generator 120, which in turn outputs electrical energy. The steam supplied to the turbines 118 is condensed back into water at condenser 122. The water from the condensed steam is recycled back to pump 110. The above process repeats itself (for a number of cycles) to generate electricity from the BWR.
A typical core of the reactor vessel may contain anywhere from about 200 to about 900 of fuel bundles B. Of course, different configurations as well as different numbers of fuel bundles may be used, so long as such configurations satisfy the safety and energy output requirements of a nuclear reactor. As shown in FIG. 2, a given fuel bundle B includes an outer channel C surrounding a plurality of fuel rods 100 extending generally parallel to one another between an upper plate U and a lower tie plate L. In general, the fuel rods 100 are provided in a generally rectilinear matrix as illustrated in FIG. 3.
The fuel rods 100 are maintained laterally spaced from one another by a plurality of spacers S which are vertically separated from one other along the length of the fuel rods 100 within the channel C. FIG. 3 illustrates a 10×10 array of fuel rods 100 surrounded by the fuel channel C. The fuel rods 100 are arranged in orthogonally related rows to surround one or more water rods 130, with two water rods 130 shown in FIG. 3. Other configurations may be used. Reactor coolant (other than the water in the water rods) flows between fuel rods 100 and collects the heat generated from nuclear reactions occurring within the fuel rod(s) of the fuel bundle(s).
In FIG. 3, a given fuel bundle B is arranged in one quadrant of a cruciform control blade 132 (e.g., only one of four quadrants is shown in this example); a cruciform control blade is a conventionally-known control blade configuration for a BWR. A given fuel bundle B is typically arranged in each of the other three quadrants of the control blade 132, as is known. Movement of control blade(s) 132 up (and down) between the bundles B controls the amount of nuclear reactivity occurring in the bundles B. FIGS. 1-3 illustrate just one conventional arrangement of rods 100 within a bundle of a reactor core of a BWR; other arrangements may be used as is evident to one skilled in the art.
In a given fuel rod 100, the fuel rod 100 is typically filled along various locations within its vertical span in a bundle with uranium (e.g., pellets containing the isotopes 238U and 235U), where the amount of 235U may be enriched (as desired) to account for safe operating conditions in a nuclear reactor. The isotope 235U is naturally found in uranium at a concentration of 0.711% (by weight) with the remainder of the uranium being the isotope 238U. Accordingly, as used hereafter, 0.71 represents the concentration of natural uranium.
After going through a process of enriching, the enriched uranium may contain from about 2% (or just above 2% by weight) to about 5% (by weight) 235U with the remainder of the uranium being isotope 238U. Although amounts greater than about 5% (by weight) of 235U in enriched uranium could be used, commercial power-producing nuclear reactors have traditionally been limited to 5% enriched uranium fuels.
As discussed above, fuel rod 100 may be filled with pellets of natural uranium in some parts of rod 100 and pellets of enriched uranium in other parts of rod 100, along the vertical (axial) height of the fuel rod 100. Some of the pellets may contain only natural uranium while other pellets may contain a combination of natural uranium and enriched uranium. Additionally, some pellets may contain only enriched uranium (a concentration of 235U above that found in natural uranium, i.e., >0.71). Typically, the uranium pellets may be about a half inch in height.
A typical fuel rod may 100 contain up to about 240 pellets. Also, in some parts, the fuel rod 100 may contain no pellets providing a void space (designated as V as further described herein) or the fuel rod 100 may have a truncated height (designated as E as further described herein). Thus, given fuel rod(s) 100 may be shorter than other fuel rods within a given fuel rod bundle B.
FIG. 4 is a representation of various axial sections of a given fuel rod 100. With reference to bundle B in FIGS. 2 and 3, the 10×10 array of fuel rods 100 and water rods 130 can be represented by a x-y map distribution (or radial lattice) cutting across a (e.g., horizontal) cross-section of fuel bundle B. In FIG. 4, a typical fuel rod 100 has a height of about 150 inches. FIG. 4 illustrates seven x-y map distributions along the vertical height of a rod 100, from 0 to 150 inches. Each 6-inch segment of a fuel rod 100 may be referred to as a node; thus there are 25 nodes in a 150 inch rod, with node 1 representing the bottom 6 inches of rod 100 and node 25 representing the segment between 144 to 150 inches of fuel rod 100 from its bottom.
These x-y map distributions, starting from the bottom of the fuel rod 100, are identified in FIG. 4 as the following:
(a) 26868 (extending from height=0 inches to 6 inches, node 1);
(b) 26869 (extending from height=6 inches to 54 inches, nodes 2 to 9);
(c) 26870 (extending from height=54 inches to 84 inches, nodes 10 to 14);
(d) 26871 (extending from height=84 inches to 96 inches, nodes 15 and 16);
(e) 26872 (extending from height=96 inches to 138 inches, nodes 17 to 23);
(f) 26873 (extending from height=138 inches to 144 inches, node 24); and
(g) 26874 (extending from height=144 inches to 150 inches, node 25).
FIGS. 5-11 illustrate x-y map distributions corresponding to (a) through (g) above for all fuel rods 100 of a given fuel rod bundle B at each of the seven specified locations between 0 to 150 inches of fuel rod height. The x-y map distributions of FIGS. 5-11 should be read in conjunction with FIG. 4. In FIGS. 5-11, a cell (fuel rod) with a “V” indicates that a fuel rod 100 does not exist at all at the cell position depicted in the corresponding x-y map distribution, and an “E” indicates that while the shell of a fuel rod 100 is present, the shell (of fuel rod 100) is empty at the corresponding cell position depicted in the corresponding x-y map distribution. FIGS. 5-11 should be referred to for the following discussion.
FIG. 5 illustrates the x-y map distribution (e.g., 26868) for all fuel rods of a given fuel rod bundle at a location between height=0 to 6 inches. In FIG. 5, the number “0.71” reflects the concentration (in percent by weight) of the amount of 235U present in the fuel rod at the corresponding cell position (A1-J10) between height=0 inches and 6 inches, or in other words, rods 100 at that location or node (node 1) which have natural uranium. In FIG. 5, the cells D6, D7, E6 and E7 constitute a water rod (WR) corresponding to the circle identified as 130 in FIG. 3, with the second water rod 130 of FIG. 3 denoted by cells F4, F5, G4 and G5.
As all fuel rods 100 at node 1 (0 to 6 inches from bottom) have a natural uranium concentration at that location, the x-y map distribution of FIG. 5 thus shows the formation of an all “natural” blanket layer at the bottom of the all fuel rods in a given bundle B at node 1. This 6-inch natural blanket at the bottom of a fuel bundle B is provided to help ensure safe nuclear reactor operation within specified safety and/or operating limits. FIGS. 6-9 show x-y map distributions corresponding to positions or nodes indicated in (b) through (e) above, and are provided merely for comparative reference.
FIG. 10 is an x-y map distribution (e.g., 26873) of the composition of a fuel rod bundle at a location between height=138 (or >138) to 144 inches; and FIG. 11 is an x-y map distribution (e.g., 26874) of the composition of a fuel rod bundle at a location between height=144 (or >144) to 150 inches. Conventionally in a BWR, another all “natural” blanket layer is provided at the top of the fuel rods 100 within a given bundle B at a height between 138 and 150 inches as reflected in FIGS. 10 and 11 (x-y map distributions 26873 and 26874). This 12-inch natural blanket at nodes 24 and 25 is also typically provided to ensure safe nuclear reactor operation within specified safety limits.
The use of a natural blanket at the bottom node 1 (6 inches of natural uranium) and at the top nodes (either a 6 inch blanket at node 25 (top) or a 12-inch blanket at nodes 24 and 25) is the conventional design choice for plant designers. The use of these natural blankets allows an overall reduction in bundle enrichment by reducing neutron leakage from the top and bottom of the core. Yet the larger 12-inch blanket at the top results in an effectively shorter fuel bundle, which can reduce thermal margins in non-peripheral portions of the core and hence lower thermal output.
Moreover, in addition to being able to satisfy limits for thermal parameters such as MFLPD (Maximum Fraction of Limiting Power Density), MAPRAT (the ratio of MAPLHGR or Maximum Average Planar Linear Heat Generation compared to its limit), MFLCPR (Maximum Fraction of Limiting Critical Power Ratio), and limits for reactivity parameters (cold shutdown margin (CSDM) and hot excess reactivity (HOTX)), use of the larger 12-inch blanket may complicate satisfying the limit or condition for the reactivity parameter known as the maximum subcritical banked withdrawal position (MSBWP).